With the advent of the liquid-metal-cooled fast breeder reactor and similar technologies has come a renewed emphasis on the successful recovery, reprocessing, and recycling of valuable fuel components of expended nuclear fuel elements. The term nuclear fuel as used herein shall include both fissile values (U.sup.233, U.sup.235, Pu.sup.239, and their compounds) and fertile values (U.sup.288, Th.sup.232 and their compounds), or mixtures of both. This is necessary because a nuclear fuel may be comprised entirely of one form or a mixture of both forms, dependent upon the particular fast reactor application and the technology employed.
In the reprocessing of nuclear fuels, the expended fuel rods from civilian or military reactors and other institutional sources are periodically withdrawn and transferred to reprocessing facilities. At such a facility, the rods are sheared to predetermined lengths and their contents dissolved in nitric acid producing a resultant aqueous solution containing fission product values, uranium, plutonium, and thorium among other materials.
This aqueous solution is treated with an organic solvent by solvent extraction techniques to separate and purify the recyclable fuel forms from fission product wastes. A detailed description of such purification processes may be found in Engineering for Nuclear Fuel Reprocessing, Justin T. Long, Gordon and Beach Science Publishers, Inc., New York, New York (1967), which is herein incorporated by reference, under such headings as the "purex", "25", "redox", and "butex" process descriptions.
Nuclear fuel reprocessing facilities generally involve the use of separate purification streams for each of the nuclear fuel components received. Once isolated and refined, the uranium, plutonium, and other nuclear fuel components are recombined in a fuels fabrication operation to produce a mixed nuclear fuel. These fuels are then loaded into fuel rods as either pellets or vibration-packed spherules.
A recent publication has described such a process wherein (U,Pu)C spherules are prepared by separate concentration and denitration procedures for hexavalent uranium and tetravalent plutonium. Bischoff, M., et al, "Sol-Gel Processes for Carbide Preparation," IAEA Panel on Sol-Gel Process for Fuel Fabrication, Vienna, May 21-24, 1973, EIR-Bericht Nr.236. The uranium is individually processed therein by conventional denitration techniques known in the art to produce an acid deficient uranium solution. A less conventional procedure is utilized to produce a plutonium solution involving addition of hydroxide form of an anion exchange resin to an individually processed plutonium nitrate stream followed by vacuum distillation to remove nitric acid and water. The resultant plutonium solution is suitable for mixing with the uranium solution and predetermined amounts of sol, urea, carbon and hexamethylenetetramine (HMTA) to create an acceptable sol-gel feed broth. The solutions and feed broth are stabilized by chilling to about 0.degree. C.
Separate purification systems have several inherent disadvantages. Such processes are expensive, create unnecessary waste streams that are difficult to treat and handle, and require a multiplicity of equipment and labor for performing almost duplicate operations. Additionally, the safeguarding of nuclear material of fissionable quality is imperative to avoid nuclear proliferation. For this reason, nuclear fuel reprocessors have proposed coprocessing and spiking the pure fuel forms with high-energy gamma emitters to make it practically suicidal to remove these materials from a hot cell. Because purified fuels may be pilfered before or during the fabrication stage, the gamma emitter should be capable of being incorporated at any stage of processing after the purified metals are isolated. Isotopes such as Zirconium 95, Cerium 144, and rare earth elements have been proposed for these purposes because of their physicochemical properties and availability.
The foregoing are all hazardous processes requiring remotely operable equipment and frequently the handling of dry, finely divided metal powders to create a "dusty" process. It is well known that such conditions necessitate frequent mechanical repairs and extensive decontamination work causing unwanted downtime and employee exposure to radiologically contaminated environments. Also serious problems regarding accountability and criticality of nuclear material in purified forms are thereby created.
A more recent approach to diversion resistant nuclear fuel reprocessing has been the direct coprocessing of mixed nuclear fuels by thermal decomposition or precipitation processes followed by conventional pellet formation procedures. While purified fuel forms of weapons-grade nuclear material are avoided, thereby limiting the likelihood of pilferage or diversion, these processes also involve "dusty" operations in the product conversion and compact fabrication stages. Thus, many of the same objections referenced above also apply to these processes.
While the objectionable characteristics of conventional powder processes in the product conversion and fuel fabrication stages can be eliminated through the use of wet processing schemes such as sol-gel technology, a successful coprocessing system that is directly operable on a product stream from a nuclear fuels reprocessing facility to produce loadable fuel forms has not been developed and reported in the literature to date. The principal difficulty in selecting such a system has been the development of a stable system, tolerant to impurities, producing homogenious solidified bodies of acceptable, but versatile particle size, suitable for reactor uses without inclusion of neutron captivating contaminants or exacting process and parameter controls. Experience for individual purified metal forms or for mixed metal sols to produce mixed metal forms indicate that the internal gelation mode of sol-gel technology is preferable over the other sol-gel modes for producing microspheres in the ratio of diameters of about 40:10:1 desirable for sphere-pac processes. Haas, P. A., et al, "Chemical Flowsheet Conditions For Internal Gelation to Prepare Urania Spheres," ORNL/TM-6850, Oak Ridge National Laboratory, Oak Ridge, Tenn. (1979), available from the National Technical Information Service, U.S. Department of Commerce, Springfield, Va.
An important characteristic, however, of internal gelation feeds is the acid or nitrate deficiency of the solution. Such feeds have heretofore been unknown with respect to coprocessed uranium and plutonium feeds because of the behavior of plutonium in dilute acid solutions involving valence instability, hydrolytic behavior and polymerization. This behavior can be partially controlled in individual, purified plutonium sol-gel processes through chemical additives, energy consumption, and additional waste treatment while performing tedious, time-consuming processes. However, it cannot be controlled in mixed solutions of uranium and plutonium because the ionic forms of these metals tend to oxidize each other to produce U.sup.+6 and Pu.sup.+3.
The principal novelty of the present invention lies in the discovery of a method for stabilizing the valence states of U.sup.+6 and Pu.sup.+4 in coprocessed solutions and for unexpectedly solving the unique behavioral problems of Pu in dilute nitric acid solutions. This is accomplished by what is thought to be the nitrate complexing of the plutonium through the formation of the hexanitrato plutonium complex by the uranium in solution with the plutonium donating a portion of its nitrate. Heretofore, it was not known that uranium would so act. The resultant coprocessed solution has unsurpassed tolerance to U, Pu, and fission product species contained therein. The method of this invention allows for nitrate-to-total metal ratios within the range of about 1.8 to 2.5 to systematically be prepared. Higher ratios are possible depending upon the ultimate U/Pu concentration ratio of the mixed fuel desired.